1. Field of the Invention
The present invention relates to a fuel rod for a light water reactor and a method for manufacturing the same.
2. Description of the Background
Recently, increasing importance has been given to nuclear power generation as a source of electrical power. In response, there has been an increased demand for the efficient operation of nuclear power plants. For nuclear power plants to operate efficiently, it is necessary to operate the light water reactors inside the plants efficiently. Improving the corrosion resistance of fuel rods would reduce the frequency of their replacement and improve the efficiency of the nuclear power plants.
A fuel rod comprise a zirconium cladding tube and UO.sub.2 fuel packed inside. As is shown in the cross section FIG. 1, a fuel rod 1 for a light water reactor is constructed as follows: UO.sub.2 fuel pellets 3 are packed one on top of another into a cladding tube 2 which comprises a zirconium alloy, such as Zircaloy-2 (JIS H4751ZrNT802D) or Zircaloy-4 (JIS H4751ZrNT804D). Both ends of the cladding tube 2 are then capped with end plugs 4 while the UO.sub.2 fuel pellets 3 are pressed with a spring 5, and are sealed by TIG welding. Beads 6 formed at the welded portions by the TIG welding are approximately 2 mm in width, and heat affected zones 7, similarly having widths of approximately 2 mm, are formed on both side areas of each bead. When the TIG welded portions are allowed to cool as they are, the heat affected zones 7 are ordinarily considered to be cooled at a cooling rate of approximately 100.degree. C./sec.
Before welding, the cladding tube 2 has a worked grain structure, while the end plugs 4 have equiaxed grain structures. The TIG welding changes the structure of the beads 6 and the heat affected zones 7. The beads 6 have a very fine acicular grain structure, which results from rapid cooling from the liquid phase. On the other hand, the heat affected zones 7 have a combined grain structure, of equiaxed grain structure 8 and acicular grain structure 9, as shown in FIG. 2, which results from rapid cooling from the .beta. (b.c.c.) phase or from a temperature within a range where .alpha. and .beta. phases coexists. The .beta. (b.c.c.) phase is a zirconium phase stable at high temperature.
Sufficient corrosion resistance for long periods is not achieved in fuel rods which use a cladding made of conventional Zircaloy-2 or Zircaloy-4. A more corrosion resistant zirconium alloy, as compared with conventional Zircaloy-2 or Zircaloy-4, has a composition which includes 0.6 to 2.0% by weight of Nb, 0.5 to 1.5% by weight of Sn, 0.05 to 0.3% by weight of Fe, and the balance being Zr and incidental impurities (hereinafter, this zirconium alloy will be referred to as a zirconium alloy containing Nb and Fe) (cf. Japanese Unexamined Patent Publications No. 61-170552, No. 1-188643, No. 2-4937, and No. 3-90527, Japanese Examined Patent Publications No. 7-6018 and No. 7-6019, and others). When a cladding tube 2 comprising this zirconium alloy containing Nb and Fe is used, the heat affected zones 7, obtained by allowing the TIG welded portions to cool, have grain structures as shown in FIG. 2. The Nb and Fe segregate at grain boundaries 10 between an equiaxed grain structure 8 and an equiaxed grain structure 8, between an equiaxed grain structure 8 and an acicular grain structure 9, and between an acicular grain structure 9 and an acicular grain structure 9. The concentrations of Nb and Fe at these boundaries are at most 3.5% and 0.6% by weight, respectively. The presence of intermetallic compounds containing Zr, Nb and Fe at the grain boundaries in the heat affected zones, however, was previously not known. Even if intermetallic compounds were present, their presence could not be confirmed since such intermetallic compounds would have had very fine grains with diameters below 0.01 .mu.m.
In general, fuel rods made from cladding tubes comprising the aforementioned zirconium alloy containing Nb and Fe, charged with UO.sub.2 fuel, have improved corrosion resistance when compared to cladding tubes comprising conventional Zircaloy-2 or Zircaloy-4. In some cases, however, the heat affected zone of these tubes have inferior corrosion resistance as compared to cladding tubes comprising Zircaloy-2 or Zircaloy-4.
The rate of corrosion of a fuel rod in a nuclear reactor may be evaluated outside the reactor by, for example, an autoclave test, which is performed at a temperature of 360.degree. C. under saturated water vapor pressure, i.e. approximately 190 atm. A sample fuel rod was constructed using a cladding tube comprising the aforementioned zirconium alloy containing Nb and Fe, the cladding tube was sealed with end plugs comprising Zircaloy-4 by TIG welding, and the sample was subjected to an autoclave test for 120 days under the above-described conditions. It was revealed that the bead portions due to welding and the heat affected zones of the end plugs, were covered with a black oxide film similar to the non-welded portions. The portions covered with this black oxide film had satisfactory corrosion resistance. A white oxide film was formed on the heat affected zones of the cladding tube comprising the zirconium alloy containing Nb and Fe. The portions where this white oxide film was formed had extremely low corrosion resistance. The reduction in corrosion resistance caused by formation of this white oxide film on the heat affected zones of the cladding tube having a thin wall leads to an extreme reduction in the life span of a fuel rod for a light water reactor.